Senior engineer, Nuclear Development Corporation, Tokai mura, Ibaraki,
Manager of Nuclear Fuel Development and Engineering SectionManager of Nuclear Fuel Engineering Section, Kobe Shipyard and Machinery Works, Mitsubishi Heavy Industries, Ltd.Nuclear Energy Systems Engineering Center, Nuclear Energy Systems Headquarters, Mitsubishi Heavy Industries, Ltd., Hyogo-ku, KobeNishi-ku, Yokohama,
Senior engineer, Nuclear Energy Systems Engineering Center, Nuclear Energy Systems Headquarters, Mitsubishi Heavy Industries, Ltd., Yokohama,
Manager of Nuclear Fuel Engineering Section, General Office of Nuclear and Fossil Power Production, The Kansai Electric Power Co., Inc., Osaka,
Pages: 21 Published: Jan 2000
High-temperature oxidation and mechanical strength under a loss-of-coolant accident (LOCA) were investigated using irradiated and unirradiated fuel cladding. Cladding from fuel rods irradiated up to 49 GWD/MTU in a Japanese commercial PWR and unirradiated cladding, most of which was preoxidized and precharged with hydrogen to simulate high burnup fuel, were subjected to the tests. High-temperature oxidation tests showed that the oxidation weight gain for irradiated cladding was equal to, or slightly lower than, that for unirradiated material. Preoxidized cladding showed less oxidation weight gain, and no effect of hydrogen absorption on oxidation behavior was observed. The mechanical tests (uniaxial strength and ductility) after a thermal sequence simulating a LOCA showed comparable behavior with that of unirradiated cladding, due to the recovery of the irradiated microstructure. These test results suggest: (1) there is no adverse effect of irradiation on the high-temperature oxidation behavior; and (2) radiation damage in cladding is eliminated during a LOCA condition.
zirconium, zirconium alloys, cladding tubes, loss-of-coolant accident, zirconium-steam reaction, thermal shock, nuclear materials, nuclear applications, radiation effects
Paper ID: STP14304S